US4689195A - Fuel assembly - Google Patents

Fuel assembly Download PDF

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Publication number
US4689195A
US4689195A US06/635,927 US63592784A US4689195A US 4689195 A US4689195 A US 4689195A US 63592784 A US63592784 A US 63592784A US 4689195 A US4689195 A US 4689195A
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United States
Prior art keywords
fuel
periphery
enrichment
array
rods
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Expired - Fee Related
Application number
US06/635,927
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English (en)
Inventor
Motoo Aoyama
Kunitoshi Kurihara
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Hitachi Ltd
Original Assignee
Hitachi Ltd
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Priority claimed from JP56072351A external-priority patent/JPS57187687A/ja
Priority claimed from JP56112179A external-priority patent/JPS5814080A/ja
Application filed by Hitachi Ltd filed Critical Hitachi Ltd
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Publication of US4689195A publication Critical patent/US4689195A/en
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
    • G21C3/30—Assemblies of a number of fuel elements in the form of a rigid unit
    • G21C3/32—Bundles of parallel pin-, rod-, or tube-shaped fuel elements
    • G21C3/326—Bundles of parallel pin-, rod-, or tube-shaped fuel elements comprising fuel elements of different composition; comprising, in addition to the fuel elements, other pin-, rod-, or tube-shaped elements, e.g. control rods, grid support rods, fertile rods, poison rods or dummy rods
    • G21C3/328—Relative disposition of the elements in the bundle lattice
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C7/00—Control of nuclear reaction
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
    • G21C3/30—Assemblies of a number of fuel elements in the form of a rigid unit
    • G21C3/32—Bundles of parallel pin-, rod-, or tube-shaped fuel elements
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors

Definitions

  • This invention relates to a fuel assembly for a reactor.
  • the fuel assembly consists of a predetermined number of fuel rods that are bundled by a support grating i.e., a channel box, in a group as a handling unit.
  • a predetermined number of fuel assemblies are set up in moderators in the reactor and the gaps between the fuel assemblies are kept so that the fuel assemblies can be individually inserted or withdrawn or control rods can be easily inserted between the fuel assemblies. Accordingly, the fuel rods located on the periphery of the fuel assemblies are surrounded by a greater number of moderators than those located at the central portion.
  • Fuel pellets having higher enrichment than the mean enrichment must be used in order to give the enrichment gradation. This reduces the maximal value of the atomic number of the fissionable material to be packed into the fuel assembly since there exists a limitation to the highest pellet enrichment at present, and reduces the derivable discharged burnup. This is not desirable for a high burnup core using high enrichment fuel pellets.
  • the present invention is directed to provide fuel assemblies that can effectively burn the fissionable material while equalizing the local output peaking factors, can keep the overall output uniform and can effectively consume the fuel resources.
  • the fuel assembly of the present invention is characterized in that the mean value of the quantity of the fissionable material loaded or packed in the fuel rod, per unit fuel rod, of the fuel rods of the outermost or periperal layer is lowered than that of the other fuel rods of the other protions of the fuel assembly.
  • loading quantity The following three methods can be used to reduce the quantity of the fissionable material load or packed per unit fuel rod, hereinafter called "loading quantity”.
  • the methods (1) and (2) that is, the constructions in which the mean density of the pellets form densification upon burnup that would otherwise occur in carrying out a high level of burnup.
  • the constructions (2) and (3) make it easier to carry out the density control. Since the method (3) reduces the diameter of the cladding tube of the fuel rod, the quantity of the moderator increases as much. Accordingly, the ratio of number of atoms (M/F) between the moderator (M) and the fissionable material (F) can be increased by a slight change.
  • FIG. 1 is a transverse sectional view of the fuel assemblies
  • FIG. 2 is a diagram showing the relation between the infinite neutron multiplication factor and the ratio M/F;
  • FIG. 3 is a diagram showing the relation between the infinite neutron multiplication factor and the ratio of the smeared density of the pellets at the central portion to the smeared density of the pellets on the periphery;
  • FIG. 4 is a diagram showing the relation between the relative fuel rod output and the above-mentioned ratio of the smeared density
  • FIG. 5 is a diagram showing the relation between the infinite neutron multiplication factor and the ratio of the mean enrichment at the central portion to the mean enrichment on the periphery;
  • FIG. 6 is a diagram showing the relation between the relative fuel rod output and the above-mentioned ratio of the mean enrichment
  • FIG. 7 is a diagram showing the relation of the control rod worth and the void coefficient versus the ratio M/F;
  • FIG. 8 is a diagram showing the relation between the value of Gadolinia worth and the ratio M/F;
  • FIG. 9 is a transverse sectional view of the fuel assembly in accordance with an embodiment of the present invention.
  • FIG. 10 is a transverse sectional view of the conventional fuel assembly for the sake of comparison.
  • FIG. 11 is a transverse sectional view of the fuel assembly in accordance with another embodiment of the present invention.
  • FIG. 12 is a transverse sectional view of the fuel assembly in accordance with another embodiment of the present invention.
  • each fuel assembly consists of a group of fuel rods 1 arranged in a rectangular or square array within a channel box 2 through which coolant of the core flows.
  • a control rod 4 or a neutron detector fitting pipe 5 is disposed outside the channel box 2.
  • the gap between the fuel assemblies 3 is kept so that the device such as the control rod 4 can be inserted.
  • the periphery of the fuel assembly is filled with the cooling water.
  • the fuel rods 1 positioned on the periphery of the fuel assembly 3 are surrounded with a greater quantity of water than those positioned at the central position of the fuel assembly 3.
  • the following nuclear heterogeneous effects (i) and (ii) occur between the periphery portion and the central portion of the fuel assembly 3.
  • the operation is effected at the point B, which is an over-deceleration range, on the periphery of the fuel assembly whereas the operation is effected at the point C, which is an insufficient deceleration range, at the central portion of the fuel assembly. Accordingly, the infinite neutron multiplication factor is greater at the central area of the fuel assembly.
  • the nuclear fission reaction occurs more vigorously on the periphery of the fuel assembly, thus elevating the local output peaking coefficient on the periphery and reducing the thermal allowance.
  • the mean value of the loading quantity of the fissionable material, per fuel rod, of the fuel rods on the periphery of the fuel assembly is made smaller than that at the central portion of the fuel assembly.
  • FIG. 3 shows the relation between the infinite neutron multiplication factor and the ratio of the smeared density of pellets at the central portion to the smeared density of pellets at the periphery when both loading quantity and mean enrichment of the fuel material are kept constant and
  • FIG. 4 shows the relation between the relative fuel rod power and the same ratio under the same condition as above.
  • FIG. 5 shows the relation between the infinite neutron multiplication factor and the ratio of the mean enrichment at the central portion to the mean enrichment at the periphery
  • FIG. 6 shows the relation between the relative fuel rod output and the abovementioned ratio, using the ratio of the mean enrichment at the central portion to that on the periphery as a parameter, respectively.
  • the infinite neutron multiplication factor of the fuel assembly in which the mean pellet density is lower on the periphery and higher at the central portion, will be compared with the infinite neutron multiplication factor of the fuel assembly which has the enrichment distribution, with reference to FIG. 2.
  • the ratio M/F on the periphery can be increased from B to B' by reducing the pellet mean density on the periphery of the fuel assembly as compared with the conventional fuel assembly.
  • the neutron absorption effect of the water becomes greater on the periphery of the fuel assembly when the ratio M/F increases, whereas the thermal neutrol utilization factor drops.
  • the infinite neutron multiplication factor slighly drops from D to D'.
  • the ratio M/F drops from C to C' at the central portion, on the contrary, when the pellet mean density is increased.
  • the infinite neutron multiplication factor drops from E to E' in FIG. 2 but since the mean neutron energy drops due to inflow of the neutrons thermalized on the periphery of the fuel assembly, the inifinite neutron multiplication factor slightly rises, on the contrary.
  • the drop of the infinite neutron multiplication factor of the fuel assemblies as a whole is limited in comparison with the fuel assembly having uniform enrichment distribution. Since the quantity of the fissionable material is reduced by reducing the pellet mean density on the periphery of the fuel assembly and since the above-mentioned effect of the infinite neutron multiplication factor is obtained, the relative fuel rod power can be remarkably reduced.
  • the infinite neutron multiplication factor of the fuel assembly having the enrichment distribution changes from D to D" in FIG. 2 on the periphery 2 because the enrichment becomes lower at the central area, on the contrary and rises from E to E" at the central area because the enrichment becomes higher.
  • the deceleration ratio (yZ s /Z a , where y is a mean value of decrement of the energy logarithm per collision and Z s (a) is the sectional area of macroscopic scattering (absorption)) becomes smaller.
  • the infinite neutron multiplication factor of the fuel assemblies as a whole can be increased more greatly by changing the pellet mean density than by changing the enrichment distribution.
  • FIGS. 7 and 8 show the relation between the ratio M/F and the value of control rod worth and between the ratio M/F and the value of Gadolinia worth, respectively.
  • the fuel assembly obtained in the abovementioned manner can reduce the quantity of fuel to be loaded on the fuel rods of the core without raising the mean enrichment of the fuel assemblies as a whole and without deteriorating the local output peaking coefficient. It is thus possible to reduce the quantity of natural uranium, the separation work unit (hereinafter abbreviated as "SWU") and the quantity to be reprocessed.
  • SWU separation work unit
  • the present invention does not exclude the method of generating the mean enrichment distribution and can use conjointly such a method.
  • the method can be used especially effectively for the fuel rods at the corners of the rectangular array of the fuel rods of the fuel assembly.
  • FIG. 9 shows the transverse section of this example.
  • the fuel rods 1 those represented by reference numerals 6 through 11 are employed.
  • Table 1 shows the enrichment and pellet mean density of each fuel rod.
  • Reference numeral 12 represents a water rod.
  • FIG. 10 shows the transverse section of the conventional fuel assembly.
  • fuel rods 13 through 17 shown in Table 2 are employed.
  • Reference numeral 18 represents a water rod.
  • the pellet mean density is 95% for all the rods.
  • Example 1 The mean enrichment is uniform in the fuel assemblies of both Example 1 and Comparative Example 1 and the enrichment at the central area in Example 1 is 0.9 times that in Comparative Example 1.
  • Table 3 shows the enrichment and pellet mean density at the central portion and on the periphery for each of Example 1 and Comparative Example 1.
  • the infinite neutron multiplication factor at the initial stage of burnup can be increased by about 0.8% by changing the enrichment ratio between the central area and on the periphery of the fuel assembly from the conventional value 1.4 to 11.
  • the derivable or discharged burnup can be extended and the assembly output or power reaches the same value as the conventional value even when the loading uranium quantity is smaller.
  • the local output peaking coefficient can be equalized substantially to the conventional value by changing the pellet mean density by 10% between the periphery and the central portion of the fuel assembly.
  • FIG. 4 shows the loading uranium quantity per unit output, the necessary natural uranium quantity, the SWU, the control rod worth and the void coefficient with those of the conventional values being 1, respectively.
  • FIG. 11 shows the transverse section of the fuel assembly of the present invention.
  • the fuel rods those represented by reference numerals 19 through 24 are employed.
  • the enrichment and pellet mean density of each fuel rod are shown in Table 5. No water is used.
  • the decrease of the void coefficient and flattening of the local output peaking coefficient are accomplished by reducing the pellet mean density on the periphery of the fuel assembly by 10% as compared with that at the central portion.
  • This arrangement makes it possible to replace the water rod, that has conventionally served for this purpose, by the fuel rod.
  • the number of fuel assemblies can be reduced by about 4% as compared with Example 1.
  • Table 6 shows the loading uranium quantity per unit output or powder, the necessary uranium quantity, the SWU, the control rod worth and the void coefficient in this example with the conventional values being 1.
  • This example shows the application of the present invention to a fuel assembly for a high burnup core using a fuel assembly of about 5 wt% enrichment.
  • the mean enrichment of the fuel assemblies becomes 5.1 w/o from the enrichment distribution shown in Table 7.
  • the enrichment of all the fuel rods is set to 5.1 wt% and the pellet mean density is such that the mean value on the periphery is lower by 15% than that at the central area.
  • the assembly output can be made the same as that in the case having the enrichment distribution shown in Table 7 without deteriorating the local output peaking factor of the fuel assembly, and the necessary natural uranium quantity per unit output, the uranium quantity and the SWU can be reduced by about 2% (in comparison with the case in which the enrichment distribution exists as shown in Table 7.
  • the present invention makes it possible to make the mean enrichment of the fuel assembly maximum. If the enrichment of all the fuel rods is set to the limit value of 5.5 wt%, the derivable burnup can be extended by about 3 GWd/st. This represents the cycle period by about 2 months.
  • Table 8 shows the loading uranium quantity per unit output or powder, the necessary natural uranium quantity and the SWU with those of the fuel assembly having the enrichment distribution of Table 7 being 1, respectively.
  • a output can be rendered flat by effectively utilizing the nuclear heterogeneity of the fuel assembly so that the uranium resources can be saved while keeping the mean enrichment of the fuel assembly and the local output peaking factor substantially equal to the conventional values. Since the loading uranium quantity can be reduced, the quantity to be reprocessed can be reduced and a core having higher safety can be realized.
  • FIG. 12 shows the transverse section of the fuel assembly.
  • Fuel rods represented by reference numerals 25 through 30 are used as the fuel rods.
  • Table 9 shows the enrichment and pellet diameter of each fuel rod.
  • the enrichment of this fuel assembly is determined to be 0.9 times the enrichment at the central portion of the conventional fuel assembly with the mean enrichment being equal to the conventional value.
  • the diameter of the fuel rods 27, 28, 29 on the periphery is reduced by 5% as compared with the diameter of the fuel rods 25, 26, 30 at the central portion (the pellet diameter is also reduced by 5%) so that the ratio of the cross-sectional area of the water or coolant passage to the cross-sectional area of the fuel pellet in any unit square on the periphery of the fuel assembly is made greater by 21% than that at the central area.
  • Table 10 shows the ratio of the mean enrichment on the periphery to the central portion and the ratio of the cross-sectional area of the coolant or water passage to the cross-sectional area of the fuel pellet in any unit square of the fuel assembly in comparison with the conventional values, respectively.
  • Table 11 shows the loading uranium quantity per unit output, the necessary natural uranium quantity, the SWU, the control rod worth and the void coefficient in this example in comparison with the conventional values that are 1, respectively.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Chemical Kinetics & Catalysis (AREA)
  • Monitoring And Testing Of Nuclear Reactors (AREA)
US06/635,927 1981-05-15 1984-07-30 Fuel assembly Expired - Fee Related US4689195A (en)

Applications Claiming Priority (4)

Application Number Priority Date Filing Date Title
JP56-72351 1981-05-15
JP56072351A JPS57187687A (en) 1981-05-15 1981-05-15 Fuel assembly
JP56-112179 1981-07-20
JP56112179A JPS5814080A (ja) 1981-07-20 1981-07-20 燃料集合体

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US06375588 Continuation 1982-05-06

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EP (1) EP0065697B1 (de)
DE (1) DE3266144D1 (de)

Cited By (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US6181762B1 (en) 1997-03-28 2001-01-30 General Electric Company Nuclear fuel bundle having different peak power limits
RU2219600C2 (ru) * 2001-10-04 2003-12-20 Открытое акционерное общество "Новосибирский завод химконцентратов" Способ снижения неравномерности распределения теплотехнических запасов в тепловыделяющих сборках активной зоны ядерного реактора (варианты)
US20070195919A1 (en) * 2003-11-27 2007-08-23 Marcel Bouffier Fuel Assembly For A Pressurized Water Nuclear Reactor Containing Plutonium-Free Enriched Uranium
US20100144903A1 (en) * 2007-05-04 2010-06-10 Cedars-Sinai Medical Center Methods of diagnosis and treatment of crohn's disease
US20130051509A1 (en) * 2011-08-31 2013-02-28 Hitachi-Ge Nuclear Energy, Ltd. Initial Core of Nuclear Reactor and Method of Loading Fuel Assemblies of Nuclear Reactor
US20180033501A1 (en) * 2016-08-01 2018-02-01 Kabushiki Kaisha Toshiba Nuclear reactor and a method of heat transfer from a core

Families Citing this family (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
DE3266144D1 (en) * 1981-05-15 1985-10-17 Hitachi Ltd Fuel assembly
JPS58135989A (ja) * 1982-02-08 1983-08-12 株式会社日立製作所 沸騰水型原子炉燃料集合体
US4708845A (en) * 1985-10-18 1987-11-24 Westinghouse Electric Corp. BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance
US4818478A (en) * 1987-12-07 1989-04-04 Westinghouse Electric Corp. BWR fuel assembly mini-bundle having interior fuel rods of reduced diameter
JP2772061B2 (ja) * 1989-09-22 1998-07-02 株式会社日立製作所 燃料集合体
SE9404497D0 (sv) * 1994-12-23 1994-12-23 Asea Atom Ab Bränslepatron med korta bränsleenheter

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GB955485A (en) * 1959-12-11 1964-04-15 Snecma Improvements in or relating to nuclear fuel elements
US3928128A (en) * 1972-07-26 1975-12-23 Siemens Ag Method for erecting and operating at least two nuclear reactors
JPS529792A (en) * 1975-07-15 1977-01-25 Hitachi Ltd Fuel assembly
US4059484A (en) * 1975-05-02 1977-11-22 Exxon Nuclear Company, Inc. Hybrid nuclear fuel assembly with reduced linear heat generation rates
JPS53109089A (en) * 1977-03-03 1978-09-22 Toshiba Corp Fuel assembly
US4229258A (en) * 1976-09-25 1980-10-21 Hitachi, Ltd. Fuel assembly
US4324615A (en) * 1978-11-29 1982-04-13 Hitachi, Ltd. Construction of nuclear reactor core
EP0051441A1 (de) * 1980-10-29 1982-05-12 Hitachi, Ltd. Kernreaktor und Brennelementbündel dafür
EP0065697A1 (de) * 1981-05-15 1982-12-01 Hitachi, Ltd. Brennstoffanordnung
US4483818A (en) * 1978-03-13 1984-11-20 Hitachi, Ltd. Fuel assembly
JPS6013283A (ja) * 1983-07-04 1985-01-23 株式会社東芝 沸騰水型原子炉

Family Cites Families (5)

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Publication number Priority date Publication date Assignee Title
FR1312193A (fr) * 1961-12-27 1962-12-14 Gen Electric Assemblage de combustible pour réacteur nucléaire
IL31145A (en) * 1967-12-15 1973-03-30 Gen Electric Nuclear reactor utilizing plutonium
US3844886A (en) * 1968-05-02 1974-10-29 Gen Electric Nuclear reactor utilizing plutonium in peripheral fuel assemblies
US3652744A (en) * 1969-11-19 1972-03-28 Atomic Energy Commission Method of making nuclear fuel elements
JPS5829877B2 (ja) * 1976-09-25 1983-06-25 株式会社日立製作所 沸騰水型原子炉の炉心

Patent Citations (11)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB955485A (en) * 1959-12-11 1964-04-15 Snecma Improvements in or relating to nuclear fuel elements
US3928128A (en) * 1972-07-26 1975-12-23 Siemens Ag Method for erecting and operating at least two nuclear reactors
US4059484A (en) * 1975-05-02 1977-11-22 Exxon Nuclear Company, Inc. Hybrid nuclear fuel assembly with reduced linear heat generation rates
JPS529792A (en) * 1975-07-15 1977-01-25 Hitachi Ltd Fuel assembly
US4229258A (en) * 1976-09-25 1980-10-21 Hitachi, Ltd. Fuel assembly
JPS53109089A (en) * 1977-03-03 1978-09-22 Toshiba Corp Fuel assembly
US4483818A (en) * 1978-03-13 1984-11-20 Hitachi, Ltd. Fuel assembly
US4324615A (en) * 1978-11-29 1982-04-13 Hitachi, Ltd. Construction of nuclear reactor core
EP0051441A1 (de) * 1980-10-29 1982-05-12 Hitachi, Ltd. Kernreaktor und Brennelementbündel dafür
EP0065697A1 (de) * 1981-05-15 1982-12-01 Hitachi, Ltd. Brennstoffanordnung
JPS6013283A (ja) * 1983-07-04 1985-01-23 株式会社東芝 沸騰水型原子炉

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* Cited by examiner, † Cited by third party
Title
"Fundamental Aspects of Nuclear Reactor Fuel Elements", 1976, p. 114, Olander.
"Mark's Standard Handbook for Mechanical Engineers", 1969, pp. 9-124, Baumeister et al.
Fundamental Aspects of Nuclear Reactor Fuel Elements , 1976, p. 114, Olander. *
Mark s Standard Handbook for Mechanical Engineers , 1969, pp. 9 124, Baumeister et al. *

Cited By (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US6181762B1 (en) 1997-03-28 2001-01-30 General Electric Company Nuclear fuel bundle having different peak power limits
RU2219600C2 (ru) * 2001-10-04 2003-12-20 Открытое акционерное общество "Новосибирский завод химконцентратов" Способ снижения неравномерности распределения теплотехнических запасов в тепловыделяющих сборках активной зоны ядерного реактора (варианты)
US20070195919A1 (en) * 2003-11-27 2007-08-23 Marcel Bouffier Fuel Assembly For A Pressurized Water Nuclear Reactor Containing Plutonium-Free Enriched Uranium
US7844025B2 (en) * 2003-11-27 2010-11-30 Areva Np Fuel assembly for a pressurized water nuclear reactor containing plutonium-free enriched uranium
US20100144903A1 (en) * 2007-05-04 2010-06-10 Cedars-Sinai Medical Center Methods of diagnosis and treatment of crohn's disease
US20130051509A1 (en) * 2011-08-31 2013-02-28 Hitachi-Ge Nuclear Energy, Ltd. Initial Core of Nuclear Reactor and Method of Loading Fuel Assemblies of Nuclear Reactor
US20180033501A1 (en) * 2016-08-01 2018-02-01 Kabushiki Kaisha Toshiba Nuclear reactor and a method of heat transfer from a core
US10692612B2 (en) * 2016-08-01 2020-06-23 Kabushiki Kaisha Toshiba Nuclear reactor and a method of heat transfer from a core

Also Published As

Publication number Publication date
EP0065697B1 (de) 1985-09-11
EP0065697A1 (de) 1982-12-01
DE3266144D1 (en) 1985-10-17

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